Development of ODS tempered martensitic steel for high burn up fuel cladding tube of SFR

Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.

2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05

Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.

Model calculation of Cr dissolution behavior of ODS ferritic steel in high-temperature flowing sodium environment

Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji

Journal of Nuclear Materials, 505, p.44 - 53, 2018/07

AA2017-0603.pdf:1.7MB

https://doi.org/10.1016/j.jnucmat.2018.03.054

A calculation model was constructed to systematically study the effects of environmental conditions (i.e. Cr concentration in sodium, test temperature, axial temperature gradient of fuel pin, and sodium flow velocity) on Cr dissolution behavior. Chromium dissolution was largely influenced by small changes in Cr concentration (i.e. chemical potential of Cr) in liquid sodium in the model calculation. Chromium concentration in sodium coolant, therefore, should be recognized as a critical parameter for the prediction and management of Cr dissolution behavior in the sodium-cooled fast reactor (SFR) core. Because the fuel column length showed no impact on dissolution behavior in the model calculation, no significant downstream effects possibly take place in the SFR fuel cladding tube due to the much shorter length compared with sodium loops in the SFR plant and the large axial temperature gradient. The calculated profile of Cr concentration along the wall-thickness direction was consistent with that measured in BOR-60 irradiation test where Cr concentration in sodium bulk flow was set at 0.07 wt ppm in the calculation.

In-situ residual stress analysis during thermal cycle of a dissimilar weld joint using neutron diffraction and IEFEM

Akita, Koichi; Shibahara, Masakazu*; Ikushima, Kazuki*; Nishikawa, Satoru*; Furukawa, Takashi*; Suzuki, Hiroshi; Harjo, S.; Kawasaki, Takuro; Vladimir, L.*

Yosetsu Gakkai Rombunshu (Internet), 35(2), p.112s - 116s, 2017/06

Numerical analysis of residual stress distribution on peening process

Ikushima, Kazuki*; Shibahara, Masakazu*; Akita, Koichi; Suzuki, Hiroshi; Morooka, Satoshi; Nishikawa, Satoru*; Furukawa, Takashi*

Welding in the World, 61(3), p.517 - 527, 2017/05

http://doi.org/10.1007/s40194-017-0430-4

In this study, first, an analysis method to predict the behaviour of residual stress distribution on shot peening process was proposed. In the proposed method, the load distribution on the collision of shots was modelled, and it was integrated with the dynamic analysis method based on the idealized explicit FEM (IEFEM). The accuracy of the proposed analysis system was confirmed by comparing the stress distribution on the collision of a single shot with the results analyzed by ABAQUS. The thermal elastic plastic analysis method using IEFEM was applied to the analysis of residual stress distribution of multi-pass welded pipe joint. The calculated residual stress distribution was compared with the measured residual stress distribution using X-ray diffraction (XRD). As a result, it was shown that the both welding residual distribution agree well with each other. Considering the calculated welding residual stress distribution, the modification of stress distribution due to shot peening was predicted by the proposed analysis system. As a result, the similar stress distribution with measurement by XRD was obtained in case that a large number of collisions are considered.

Development of CdTe pixel detectors combined with an aluminum Schottky diode sensor and photon-counting ASICs

Toyokawa, Hidenori*; Saji, Choji*; Kawase, Morihiro*; Wu, S.*; Hurukawa, Yukihito*; Kajiwara, Kentaro*; Sato, Masugu*; Hirono, Toko*; Shiro, Ayumi*; Shobu, Takahisa; et al.

Journal of Instrumentation (Internet), 12(1), p.C01044_1 - C01044_7, 2017/01

https://doi.org/10.1088/1748-0221/12/01/C01044

We have been developing CdTe pixel detectors combined with a Schottky diode sensor and photon-counting ASICs. The hybrid pixel detector was designed with a pixel size of 200 micro-meter by 200 micro-meter and an area of 19 mm by 20 mm or 38.2 mm by 40.2 mm. The photon-counting ASIC, SP8-04F10K, has a preamplifier, a shaper, 3-level window-type discriminators and a 24-bits counter in each pixel. The single-chip detector with 100 by 95 pixels successfully operated with a photon-counting mode selecting X-ray energy with the window comparator and stable operation was realized at 20C. We have performed a feasibility study for a white X-ray microbeam experiment. Laue diffraction patterns were measured during the scan of the irradiated position in a silicon steel sample. The grain boundaries were identified by using the differentials between adjacent images at each position.

Corrosion evaluation of the dissolver

Suzuki, Kazuyuki; Nakamura, Yoshinobu; Hikita, Keiichi; Kogawa, Takayuki; Hayashi, Shinichiro

no journal, , 

The reprocessing plant dissolves Spent Fuels and it is a process to attract U and Pu after the removal of the Fission Products. Because dissolver is a tank with boiling nitric acid, there is it in the severe corrosion environment. Corrosion environment of the heat transfer surface is the severest. Therefore, the dissolver is made in High Chrome Nickel Steel in consideration of corrosion resistance. This paper reports a supersonic wave measurement method and corrosion rate of the dissolver.

High temperature creep properties of ODS steel cladding for evaluating severe accident

Kato, Shoichi; Furukawa, Tomohiro; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Oka, Hiroshi; Inoue, Toshihiko; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.

no journal, , 

Oxide dispersion strengthened (ODS) steel is a prime candidate for cladding tubes of Japan Sodium-cooled Fast Reactor (JSFR) due to the high temperature and radiation resistances. One of the safety design of JSFR for Design Extension Condition (DEC) is the control of severe plant conditions, including prevention of severe accidents and mitigation of severe-accident consequences. Therefore, it is necessary to acquire the mechanical properties at ultra-high temperature conditions for core materials to evaluate safety design. There are, however, no data for ODS claddings at ultra-high temperature condition for the reflecting to the design criteria. In this study, creep rupture tests of 9Cr-ODS, 12Cr-ODS and FeCrAl-ODS steel claddings have been done at elevated temperatures, and the effect of minor elements such as Al, Zr and O on the mechanical strength and the creep rupture curve for the safety design were evaluated. The effect of minor elements was estimated based on the data at 700C and 1000C. As the results, it was confirmed that the addition of Zr had an effect on the improvement of creep strength at elevated temperature for the FeCrAl-ODS steel claddings.

Transient burst properties of ODS steel cladding for evaluating sever accident

Inoue, Toshihiko; Sekio, Yoshihiro; Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Furukawa, Tomohiro; Kaito, Takeji; Torimaru, Tadahiko*; Hayashi, Shigenari*; et al.

no journal, , 

In order to evaluate the strength and deformation in severe accident, the transient burst tests were carried out with various heating rates (from 0.1 to 10 K/s) and hoop stresses (from 50 to 200 MPa) to provide more evaluation data. The test materials were core materials in fast reactors, 9-18Cr-ODS and accident tolerant fuel cladding tube in the light water reactors, FeCrAl-added ODS ferritic steels. Result, the rupture strength dropped with increasing hoop stress and decreasing heating rate. The burst strength of Al-added ODS steels was lower than other ODS steels, Al and Zr-added ODS steels show good transient burst strength.

Modelling and numerical calculation of mass transfer phenomena between fast reactor fuel cladding tube and liquid Na

Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Uwaba, Tomoyuki; Kaito, Takeji; Furukawa, Tomohiro; Kato, Shoichi

no journal, , 

Maximum temperature of ODS steel cladding tube for long life fast reactor fuel is very high (approximately 700C) in normal operation condition. It was reported that, in reactor operation, mass transfer phenomena (dissolution, deposition, penetration) took place as a result of increased solubility of steel constituent elements in liquid Na. The driving force of these phenomena is the chemical potential gap of solute elements in steel and liquid Na, which is dependent of not only temperature but also other factors such as impurity concentrations in Liquid Na. For appropriately evaluating experimental data and predicting the corrosion behavior in actual plant, it is required to list up the key factors including other factors than temperature and residence time and understand the effects of these factors. In this study, transfer behavior of Cr (main alloying element of ODS steel) is discussed; modelling and numerical calculation were carried out on Cr dissolution behavior from fast reactor fuel cladding tube into liquid Na.

Evaluation of the sustainability of compressive residual stress by introduced a stress improvement method, 1; Development of analysis method of the shot peening using idealized explicit FEM

Yamada, Yusuke*; Shibahara, Masakazu*; Ikushima, Kazuki*; Kitani, Yuji*; Nishikawa, Satoru*; Furukawa, Takashi*; Akita, Koichi; Suzuki, Hiroshi; Morooka, Satoshi

no journal, , 

no abstracts in English